Process for solidifying radioactive wastes by addition of lime to precipitate fluoride

ABSTRACT

A PROCESS FOR SOLIDIFYING RADIOACTIVE WASTES CONTAINING AT LEAST ONE MEMBER SELECTED FROM THE GROUP CONSISTING OF WATER-SOLUBLE FLUORIDE, WATER-SOLUBLE SULPHATE AND MIXTURES THEREOF COMPRISING MIXING WITH AGITATION AT LEAST ONE LIQUID RADIOACTIVE WASTE WITH SUFFICIENT LIME TO NEUTRALIZE THE FREE SULPHURIC ACID AND IN PRECIPITATE THE FREE FLUORIDE, EVAPORATING THE WATER PRESENT IN THE MIXTURE UNTIL A HARDENABLE PASTE IS OBTAINED, HARDENING THE PASTE AND ENCLOSING THE RESULTING BLOCK IN A WATERTIGHT COVERING.

United States Patent PROCESS FOR SOLIDIFYING RADIOACTIVE WASTES BYADDITION OF LIME TO PRE- CIPITATE FLUORIDE Emile Detilleux, Werner GeorgHild, Gianfranco Lazzaretto, Emilio Manuel Menchero-Lopez, and RudolfRometsch, all of Eurochemic, Mol, Belgium N0 Drawing. Filed Nov. 18,1966, Ser. No. 595,325 Claims priority, application Belgium, Apr. 7,1966, 26,521, Patent 679,231 Int. Cl. GZlc 19/46 U.S. Cl. 252-301.1 15Claims ABSTRACT OF THE DISCLOSURE A process for solidifying radioactivewastes containing at least one member selected from the group consistingof water-soluble fluoride, water-soluble sulphate and mixtures thereofcomprising mixing with agitation at least one liquid radioactive wastewith sufficient lime to neutralize the free sulphuric acid and toprecipitate the free fluoride, evaporating the Water present in themixture until a hardenable paste is obtained, hardening the paste andenclosing the resulting block in a watertight covermg.

The present invention concerns a process for the treatment of thedecladding wastes, some medium level Wastes and the high level wastesproduced by plants reprocessing irradiated nuclear fuels.

.The invention concerns a process for the insolubilisasion andself-solidification of the eflluents considered, after evaporation ofthe excess water. This process is simpler than those previouslyenvisaged for the treatment of these particular efiluents.

The object of the invention is to provide a process for the treatment,in the same operation, of the different types of solutions whichconstitute hese categories of effiuentsa process which has not beenachieved before now.

The invention enables a solid compact mass to be obtained which lendsitself particularly well to:

(a) Either being covered by a thin layer of an insoluble material suchas bitumen,

(b) Or being placed in drums or other leak-proof containers, where thelevel of radioactivity emitted by the solid mass would be incompatiblewith the radiation resistance of the covering material considered under(a).

This covering or containment acts as a barrier against leaching of theradioactive components contained in the solid mass. In particular, thecovering of the solid considered under (a) makes it possible to avoiddispersing the radioactive solids in an inert mass which serves at thesame time as 'binder and fixing agent preventing leaching. The processaccording to the invention also achieves a large volume reduction of thesolids to be stored compared With that achieved by the known dispersionmethods, proposed notably for slurries with specific activity inferiorto that considered here, since in these known techniques the volume ofactive material incorporated in the inert mass represents only 40-50% ofthe total volume to be stored.

The process according to the invention does not require any operation athigh temperature; this considerably reduces the risks of acceleratedcorrosion due to the chemical nature of the solutions treated, as wellas the risks of volatilising certain radioactive species. It ispossible, therefore, to make use of the present-day constructionmaterials and to avoid complex and delicate installations for thepurification of gases and vapours produced by the operation. Because ofthis, the process according to the invention represents a notablesimplification compared with other techniques which have been envisaged,such as calcination of the efiiuents or their incorporation in vitrifiedmasses.

It is well known that nuclear fuels are essentially made up of a can ofa weakly neutron-absorbing material, generally aluminium, magnesium,stainless steel or zirconium and its alloys such as Zircaloy, and of anuranium core in the form of metal or oxide. After irradiation in areactor, the core contains some plutonium and fission products inaddition.

The technique of processing irradiated fuels is well known. Theprincipal steps are as follows:

(1) Removal of the canning, either mechanically or chemically.

(2) Dissolution of the core in nitric acid.

(3) Separation of the uranium, plutonium and fission products by aseries of extractions using selective agents such as tributyl phosphate.

Chemical decladding uses selective agents for dissolving the canning;these are caustic soda for aluminium, sulphuric acid for magnesium orstainless steel, and ammonium fluoride for zirconium or Zircaloy. Thesedecladding solutions constitute a very large part of the volume ofmedium level wastes produced by a reprocessing plant, as they contributefrom 0.5 to 10 m. per ton of uranium treated. The activity contained inthese solutions arises partly from the fission products which passedinto the solution in the course of decladding, and partly from theactivation of certain elements contained in the canning material. Theactivity level measured in these solutions can reach 10 Ci/m. dependingon the irradiation conditions and the cooling time of the fuelprocessed.

As an illustration, the average composition of the decladding solutionsis given in the table hereafter:

Canning Average-composition Aluminium:

NaOH 2.5-3 M NaAlO 1 1.5-3 M Magnesium:

MgSO, 1.25-1.75 M Stainless steel:

(SS)SO 0.8-1 M Zirconium:

NH F 0.5-0.6 M

NH NO 0.08-0.1 M Zircaloy:

(NHQ ZrF 0.55-0.65 M

-80 g./l. A1.

2 30-42 g./l. Mg.

5 -55 g./1. SS.

4 -58 g./1. Zr.

As no simple method for the treatment of these efiluents has beenproposed up to now, considerable research has been done into mechanicalmethods to take the place of chemical decladding. This research has ledto complex, delicate and onerous techniques.

The other medium level wastes considered in the process according to theinvention are roduced essentially in the course of the third stage ofreprocessing. They are made up of the concentrates resulting from theevaporation of different wastes of lower specific activity. Theircomposition is rather variable around the following values:

Main component: NaNO lO400 g./l.

Traces of: N0 SO; CO H PO U, other metals, Si, tributyl, dibutyl andmonobutyl phosphates, kerosene.

Activity: 10 -10 Ci/m.

Quantity: 5 m. /tonne U processed.

The large volumes of these types of Waste, their relatively highradioactivity, their corrosive nature and their chemical compositionwhich prevents their mixing, make prolonged liquid storage in separatetanks unacceptable from an economic point of view. Therefore, atreatment giving a drastic volume reduction and a transformation to aconvenient form for prolonged, economic and safe storage is highlydesirable.

The high level wastes are also produced during the third stage of thereprocessing. They contain essentially the bulk of the fission productsas well as a certain number of other salts originating in the reagentsadded during processing and from corrosion of the equipment. At the endof the process, they are usually concentrated by evaporation. These highlevel wastes have variable compositions, which can be summarized asfollows:

Fission and corrosion products: $100 g./l. Compounds originating in thereagents: 120 g./l. HNO 11 M 5+) activity: Ci/m. Quantity: 1 m. /t.

1 Estimated as oxides.

Until now, no special treatment for the decladding waste has beenproposed for resolving this problem in a simple manner. However, sometreatment processes applied to very high level wastes (concentratedfission products, very low level wastes, and medium level wastesexcluding decladding waste solutions are known or have been studied.

The most important of these known processes are:

(a) various calcination processes,

(b) incorporation in glass,

(c) disposal in deep wells,

(d) co-precipitation, flocculation, ion exchange, foam separation, etc.

(e) incorporation in bitumen, cement, polymerising resins,

paraflin, waxes, etc.

Although none of these methods has so far been adapted and used fordecladding waste solutions, it is evident from the nature of theeffluents and concentrates considered here that the application of theprocesses mentioned would lead to certain technical difficulties as wellas very high costs.

Processes (a) and (b) present very serious technological and economicdrawbacks due to the quantity of material to be treated; the saltcontent on the extremely corrosive nature of the solutions, thevolatilsation of certain fission products, the operating temperaturesand the risks of explosion involved.

Method (c) is applicable only in regions with a particular geologicalstructure, and only a limited number of these exist.

The methods listed under (d) are uneconomic in view of the large volumesto be treated, and also because of the corrosion risks, content andvariety of salts, which would necessitate the use of different processesappropriate to each case.

The techniques classed under (e) involve fire and explosion risks, aswell as risks of radiation damage to the 4 sheathing material, and aboveall the volume reduction is insufficient.

The research carried out by the applicant into a method particularlysuited to decladding wastes has led to surprising results which not onlyenable them to be solidified by a simple process but also enable othercategories of radioactive wastes to be incorporated into the solidobtained.

The invention concerns a process for the solidification of radioactivewastes containing either water-soluble fluoride or a water-solublemixture of fluoride and sulphate, possibly in the presence of a maximumof about 1 molar equivalent of nitrate for 3 molar equivalents offluoride and/or sulphate. The process is characterized:

By the simultaneous mixing and agitation of at least one liquid eflluentwith lime, the quantity of lime used being at least equal to thatnecessary to neutralize the free sulphuric acid and/ or to precipitatethe free fluoride,

By the subsequent evaporation of the excess water present in the mixtureuntil a slurry is formed which hardens on standing,

By the hardening of this slurry,

And by the sealing of the cake obtained in a watertight containment.

The first stage of the process according to the invention consists ofprecipitating the fluorides and sulphates from the decladding solutionswith lime according to well known techniques. The normal treatment toapply would be the separation of the fluoride and sulphate precipitate,by filtration for example. This would lead to a powdery productdiflicult to handle, from which the radioactive components could beeasily leached. In addition, a large quantity of the radioactivesubstances would be left in the large volume of mother liquor, whichwould require a subsequent treatment.

It has now been found that if, instead of the separa tion of theprecipitate by one of the usual methods (filtration, decantation, etc.),the bulk of the water is evaporated, a slurry results which contains 10to 20% by weight of water, is easy to handle and can be freely pouredinto any receptacle. On cooling and standing for several hours in air,the slurry hardens and enclose all the radioactive compounds.

The second and third stages of the process according to the invenutiontherefore consist of distillation or evaporationof the water from themixture until a paste containing 10 to 20% by weight of water isobtained, which is then allowed to harden to a dry cake.

The leachability of the cake obtained by this process is then reduced toa very small amount (less than 2.10- g./cm. day) by covering the hardcake with an impermeable and insoluble material, which acts as a barrierbetween the cake containing the radioactive salts and the environment.This covering is the fourth step of the process according to theinvention. An effective coating can be obtained by:

Immersing the dried cake successively into two baths of molten bitumen,one of low viscosity (170190 C.), and the other of higher viscosity (l50C.),

Placing the dry block in a bitumen-lined container of metal or concrete,and completely filling the free space with molten bitumen. The previouscoating of the internal walls may be omitted.

Covering materials other than bitumen, for example hardenable resins,cement, etc. can be considered.

In cases where the use of an impermeable, insoluble covering materialwould be incompatible with the level of radiation emitted by the block,the block can be placed in a leak-tight metal cask of corrosionresistant alloy. In this case, the risk of leaching the block byexternal agents is non-existant.

The principal advantage of the process according to the invention, apartfrom its simplicity and low cost, lies in the fact that the phenomenonof solidification of the concentrated slurry takes place over a largerange of waste compositions to be treated. This property is quiteunforeseen, particularly for solutions with a low sulphate content. Themechanism of solidification of a slurry obtained by the processdescribed in the present invention can very well be explained forsolutions containing only sulphates, because it is similar to thesolidification of plaster of paris, although the preparation of this iscompletely different.

The innovations of the present invention lie in the fact that it is notnecessary to pass through a stage of forced drying as in the preparationof plaster of paris, and that it is possible, at the same time, tosolify fluoride solutions from the dissolution of zirconium andZircaloy, although it has been established that calcium fluoridecrystallizes with no molecules of water of crystallization. In addition,the solidification process is not disturbed by the presence of solublesalts, such as MgSO NaNO etc. This allows great latitude in the choiceof volumes of solutions for solidification, as is shown, for example, bythe following table:

Relative volumes Minimum Maximum 1 Or 0. 1 Or 1.

The incorporation and the solidification of the high level wastes iscarried out either after neutralization of the free nitric acid byCa(OH) CaO, NaOH, etc. or after partial destruction of this acid by areducing agent, such as sugar or formaldehyde, followed byneutralization of the remaining acid. This latter method leads to areduction in volume to be treated.

This process according to the present invention can be applied to thefinal treatment and storage of radioactive waste solutions, andparticularly to the treatment of decladding waste solutions, since thesolid cakes obtained, suitably covered with an impermeable, insolublematerial and placed in an appropriate containment, such as inexpensivecasks, either can be stored in the open with no protection than anenclosure preventing entrance to the radiation zone, or can be throwninto the sea. In this case, the long-term storage costs are reduced to aminimum. For high level wastes, storage should probably be carried outfor a certain period under shelter with forced ventilation to ensure thedissipation of the heat released by fission product decay.

Some ways of carrying out the process according to the invention aredescribed in the following examples:

EXAMPLE 1 265 g. of lime (technical grade containing about 90% Ca(OH)are added in -20 minutes to 1.500 cm. of Zircaloy decladding wastesolution with a specific activity of 1 Ci/l. and mixed by a radial flowturbine type agitator. After distillation of 1.150 cm. of water, theslurry obtained is cooled in a detachable cylindrical mould. A liftinghook is placed in the viscous mass, which is allowed to cool in air.After one hour, the paste has changed into a hard cake. After thecylindrical mould has been dismantled, the block is plunged, by means ofthe hook, into two successive baths of hot bitumen, the first at 180 C.,the second at 130 C.

The coated block is then placed in a bath of distilled water keptbetween 21 C. and 30 C. for a period of days, the :water being renewedevery 12 hours. The 80 litres of water used in this test areconcentrated to 1 litre, and no activity is detected in the water.

The apparent density of the block, after it has been kept for 2 hours at110 C. is 1.5 g./cm. and the volume reduction obtained (volume ofinitial solution/ volume of block) is 3.3.

EXAMPLE 2 133 g. of technical grade lime are mixed in 10 to 20 minuteswith 500 cm. stainless steel decladding solution containing 3 M free H50 and 5 Ci radioactivity/l, in the way described in Example 1. 350 cm.water are distilled and the resulting slurry treated as described inExample 1. The paste hardens in 45 minutes. No activity is detected inthe water used for the leach tests. The apparent density of the block is1.5 g./cm. and the volume reduction obtained is about 2.

EXAMPLE 3 EXAMPLE 4 500 cm. magnesium decladding solution with aspecific activity of 300 mCi/l. are treated with 46 g. of technicalgrade lime as in the previous examples. After evaporation of 425 cm. ofwater, the paste hardens in 50 minutes. No activity is detected in thewater used for leach tests on the block after it has been coated withbitumen. The apparent density of the block is 1.4 g./cm. and thecorresponding reduction is 4.

EXAMPLE 5 500 cm. magnesium decladding solution identical to that ofExample 4, are treated with 26 g. technical grade lime. Then 468 cm.aluminum decladding solution iden tical to that used in Example 3 areadded. After evaporation of 700 cm. water, the paste hardens in 75minutes. No activity is detected in the water used for leach tests. Theapparent density of the block is 1.2 g./cm. and the corresponding volumereduction is about 4.5.

EXAMPLE 6 250 cm. stainless steel decladding solution (see Example 2)are mixed with 250 cm. magnesium decladding solution (see Example 4),the neutralized with g. technical grade lime. After evaporation of 350cm. water, the paste hardens in 1 hour. No activity is recorded in theleach test water. The apparent density of the block is 2.2 g./cm. andthe volume reduction is 4.3.

EXAMPLE 7 250 cm. stainless steel decladding solution (see Example 2)mixed with 250 cm. decladding solution (see Example 4) are neutralizedwith 80 g. technical grade lime. Then 234 cm. aluminum decladdingsolution (see Example 3) are added. After evaporation of 550 cm. water,the paste hardens in 80 minutes. No activity is detected in the leachtest water. The apparent density of the block is 2, and thecorresponding volume reduction is 3.3.

EXAMPLE 8 350 cm. stainless steel decladding solution (see Example 2),after neutralization With 80 g. technical grade lime, are mixed with1.500 cm. Zircaloy decladding solution (see Example 1). Afterevaporation of 1.570 cm. water, the paste hardens in 80 minutes. Theactivity detected in the water from the leach tests corresponds to 10ppm. of the total activity involved in the test. The apparent density ofthe block is 1.6 g./cm. and the volume reduction is 6.5.

7 EXAMPLE 9 550 cm. stainless steel decladding solution (see Example 2),after neutralization by 138 g. technical grade lime, are mixed with 300cm. aluminum decladding solution (see Example 3) and 1.500 cm. Zircaloydecladding solution. After distillation of 2.000 cm. water, the pastehardens in 90 minutes. The activity recorded in the water from the leachtests corresponds to 25 p.p.m. of the total activity involved in thetest. The apparent density of the block is 2 g./cm. and the volumereduction is 7.4.

EXAMPLE 500 cm. stainless steel decladding solution (see Example 2)mixed with 50 cm. magnesium decladding solution (see Example 4) areneutralized with 125 g. technical grade lime. Then 50 cm. aluminumdecladding solution (see Example 3) and 1.500 cm. Zircaloy decladdingsolution (see Example 1) are added. After evaporation of 1.735 cm.water, the paste hardens in 90 minutes. The activity detected in thewater from the leach tests corresponds to 20 p.p.m. of the totalactivity involved in the test. The apparent density of the block is 1.7g./ 0111. and the volume reduction is 6.

EXAMPLE 1 1 250 cm. stainless steel decladding solution (see Example 2)added to 50 cm. magnesium decladding solution (see Example 4) areneutralized by 72 g. technical grade lime. After addition of 856 cm.concentrate with a specific activity of 10 Ci/l. and 300 cm. Zircaloydecladding solution (see Example 1), 1.100 cm. water are distilled off.The paste hardens in 130 minutes. The activity detected in the waterfrom the leach tests corresponds to 135 p.p.m. of the total activityincloved in the test. The apparent density of the block is 2.1, and thevolume reduction is 4.7.

EXAMPLE 12 250 cm. stainless steel decladding solution (see Example 2),added to 50 cm. magnesium decladding solution (see Example 4), areneutralized by 72 g. technical grade lime. After addition of 120 cm.high level waste (100 ci/l.) previously partially neutralized by 43 g.technical grade lime and 300 cm. Zircaloy decladding solution (seeExample 1), 550 cm. water are distilled off. The paste hardens in 20minutes. The activity detected in the water from the leach testscorresponds to 85 p.p.m. of the total activity. The apparent density ofthe block is 1.8 g./cm. and the corresponding volume reduction is 3.3.

EXAMPLE 13 250 cm. stainless steel decladding solution (see Example 2)and 50 cm. magnesium decladding solution (see Example 4) are neutralizedby 72 g. technical grade lime. After addition of 856 cm. of concentrate(see Example 11) and 300 cm. Zircaloy decladding solution (see Example1), 120 cm. high level waste are added (see Example 12), the acidity ofwhich, previously reduced to 0.5 M HNO by treatment with sugar, has beencompletely neutralized by 3 g. technical grade lime. After distillationof 1.250 cm. water, the resulting paste hardens in 65 minutes. Theactivity detected in the water from the leach tests corresponds to 115p.p.m. of the total. The apparent density of the block is 1.9, and thevolume reduction is 4.6.

EXAMPLE 14 250 cm. high level waste solution (see Example 13) mixed with125 cm. 12 M sulphuric acid are reduced by distillation to 250 cm. Tothis concentrate is added 250 cm. stainless steel decladding solution(see Example 2) and the whole is neutralized by 176 g. technical gradelime. After addition of 300 cm. Zircaloy decladding solution (seeExample 1), 600 cm. water are distilled off. The paste hardens in 35minutes. The activity detected in the leach test water corresponds top.p.m. of the total. The apparent density of the block is 2 g./cm. andthe volume reduction is 4.9.

EXAMPLE 15 cm. 12 M sulphuric acid are added to 500 cm. high level wastesolution (see Example 13). The mixture is evaporated down to a volume of250 cm. Then 80 g. technical grade lime are added. 175 cm. water aredistilled off. The paste hardens in 30 minutes. The ac tivity detectedin the leach water corresponds to p.p.m. of the total. The apparentdensity of the block is 1.7 g./cm. and the volume reduction is about 4.

What we claim is:

1. A process for solidifying radioactive wastes containing at least onemember selected from the group consisting of water-soluble fluoride,water-soluble sulphate and mixtures thereof comprising mixing withagitation at least one liquid radioactive waste with sufficient lime toneutralize the free sulphuric acid and to precipitate the free fluoride,evaporating the water present in the mixture until a hardena'ble pasteis obtained, hardening said paste and enclosing the resulting block in awater-tight covering.

2. A process according to claim 1, wherein sufiicient water isevaporated to obtain a paste containing from about 10 to about 20% byweight of Water.

3. A process according to claim 1 wherein the waste is obtained from thedecladding of irradiated nuclear fuels.

4. A process according to claim 3, wherein a medium level waste is addedto the decladding waste said medium level waste having an activity offrom about 10 to 4 times 10 Ci/rn.

5. A process according to claim 3, wherein to the decladding waste hasbeen added a high level waste, the free acid contained therein havingbeen neutralized beforehand, said high level waste having an activity ofmore than 3 times 10 Ci/m.

6. A process according to claim 1 wherein a high level waste isemployed, the free acid contained therein having been neutralizedbeforehand, and to which fluoride and sulphate has been added before theaddition of lime, said high level waste having an activity of more than3 times 10 Ci/m.

7. A process according to claim 1, wherein the block is covered with animpermeable material, insoluble in water.

8. A process according to claim 7, wherein said impermeable material isselected from the grou consisting of bitumens, asphalts, parafiins, andhardenable resins.

9. A process according to claim 1, wherein the paste is placed in acorrosion-resistant metal container after the water has been evaporated,allowed to harden, and the container is sealed with the hardened blockinside.

10. A process according to claim 1 wherein said waste contains nitratein a concentration of about 3 molar equivalents per molar equivalent offluoride and sulphate.

11. A process according to claim 2, wherein the waste ifs 1 btained fromthe decladding of irradiated nuclear ue s.

12. A process according to claim 11, wherein medium level waste is addedto the decladding waste, said medium level waste having an activity offrom about 10 to about 4 times 10 Ci/m.

13. A process according to claim 11, wherein to the decladding waste hasbeen added a high level waste, the free acid contained therein havingbeen neutralized beforehand, said high level waste having an activity ofat least 3 times 10 Ci/m.

14. A process according to claim 1, wherein said block is laced in acontainer.

15. A process according to claim 14, wherein said 9 10 block is firstcoated with an impermeable, insoluble 3,330,771 7/1967 Komatsv et a1252-3011 material. 3,332,884 7/1967 Kelmar 252301.1

References Cited UNITED STATES PATENTS CARL D, QUARFORTH, Pnmary Exammer3,272,756 9 1966 Kasey 5 301 5 F-M-GITTEiASSiStantEXaminer 3,274,7849/1966 Shock etal. 252301.1X

